FeCrAl fuel/clad chemical interaction in light water reactor environments

Haozheng J. Qu; Maria Higgins; Hamdy Abouelella; Fabiola Cappia; Jatuporn Burns; Lingfeng He; Caleb Massey; Jason Harp; Kevin G. Field; Richard Howard; Rajnikant V. Umretiya; Andrew K. Hoffman; Janelle P. Wharry; Raul B. Rebak

doi:10.1016/j.jnucmat.2023.154717

FeCrAl fuel/clad chemical interaction in light water reactor environments

Haozheng J. Qu
, Maria Higgins
, Hamdy Abouelella
, Fabiola Cappia
, Jatuporn Burns
, Lingfeng He
, Caleb Massey
, Jason Harp
, Kevin G. Field
, Richard Howard
, Rajnikant V. Umretiya
, Andrew K. Hoffman
, Janelle P. Wharry
, Raul B. Rebak

Nuclear Materials Performance

Research output: Contribution to journal › Article › peer-review

20 Scopus citations

Abstract

This article investigates the fuel-cladding chemical interaction (FCCI) behavior of two commercial FeCrAl alloys, APMT composition (Fe-21Cr-5Al-3Mo wt.%) and C35M (Fe-13Cr-5Al-2Mo-0.2Si-0.03Y wt.%), after neutron irradiation. “H-cup” diffusion multiples of FeCrAl alloys and ceramic UO₂ fuel were irradiated at a temperature of ∼300 °C to a total estimated burnup of 26 GWd/tHM. Post-irradiation Examination results demonstrate the excellent degradation resistance of FeCrAl alloys as accident tolerant fuel (ATF) cladding materials in light water reactor conditions. The study concludes that there was no irradiation-induced defects observed in either of the two commercial FeCrAl claddings. The formation of amorphous Al/U mixed oxide was observed at the fuel-clad interface, which can serve as a tritium permeation barrier and protect against potential chemical attack from the fuel. The study attributed the formation of amorphous Al/U mixed oxide to the low temperature and limited time of neutron irradiation. APMT forms more distinct Cr and Cr-Fe intermetallic at the FeCrAl-UO₂ interface than C35M due to the higher bulk Cr:Al ratio.

Original language	English
Article number	154717
Journal	Journal of Nuclear Materials
Volume	587
Early online date	Sep 8 2023
DOIs	https://doi.org/10.1016/j.jnucmat.2023.154717
State	Published - Dec 15 2023

Keywords

Accident tolerant fuel cladding
FeCrAl alloy
Fuel clad chemical interaction

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10.1016/j.jnucmat.2023.154717

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@article{9f99ace87fad4581a94803e188dd5cbe,

title = "FeCrAl fuel/clad chemical interaction in light water reactor environments",

abstract = "This article investigates the fuel-cladding chemical interaction (FCCI) behavior of two commercial FeCrAl alloys, APMT composition (Fe-21Cr-5Al-3Mo wt.\%) and C35M (Fe-13Cr-5Al-2Mo-0.2Si-0.03Y wt.\%), after neutron irradiation. “H-cup” diffusion multiples of FeCrAl alloys and ceramic UO2 fuel were irradiated at a temperature of ∼300 °C to a total estimated burnup of 26 GWd/tHM. Post-irradiation Examination results demonstrate the excellent degradation resistance of FeCrAl alloys as accident tolerant fuel (ATF) cladding materials in light water reactor conditions. The study concludes that there was no irradiation-induced defects observed in either of the two commercial FeCrAl claddings. The formation of amorphous Al/U mixed oxide was observed at the fuel-clad interface, which can serve as a tritium permeation barrier and protect against potential chemical attack from the fuel. The study attributed the formation of amorphous Al/U mixed oxide to the low temperature and limited time of neutron irradiation. APMT forms more distinct Cr and Cr-Fe intermetallic at the FeCrAl-UO2 interface than C35M due to the higher bulk Cr:Al ratio.",

keywords = "Accident tolerant fuel cladding, FeCrAl alloy, Fuel clad chemical interaction",

author = "Qu, \{Haozheng J.\} and Maria Higgins and Hamdy Abouelella and Fabiola Cappia and Jatuporn Burns and Lingfeng He and Caleb Massey and Jason Harp and Field, \{Kevin G.\} and Richard Howard and Umretiya, \{Rajnikant V.\} and Hoffman, \{Andrew K.\} and Wharry, \{Janelle P.\} and Rebak, \{Raul B.\}",

note = "Funding Information: The financial support of Global Nuclear Fuel Americas, GE Hitachi Nuclear Energy, and GE Research is gratefully acknowledged. This material is based upon work supported by the Department of Energy under Award Number DE-NE0009047 , and the Department of Energy-Nuclear Science User Facilities program under award NSUF-RTE-20-4108 . Publisher Copyright: {\textcopyright} 2023",

year = "2023",

month = dec,

day = "15",

doi = "10.1016/j.jnucmat.2023.154717",

language = "English",

volume = "587",

journal = "Journal of Nuclear Materials",

issn = "0022-3115",

publisher = "Elsevier B.V.",

}

TY - JOUR

T1 - FeCrAl fuel/clad chemical interaction in light water reactor environments

AU - Qu, Haozheng J.

AU - Higgins, Maria

AU - Abouelella, Hamdy

AU - Cappia, Fabiola

AU - Burns, Jatuporn

AU - He, Lingfeng

AU - Massey, Caleb

AU - Harp, Jason

AU - Field, Kevin G.

AU - Howard, Richard

AU - Umretiya, Rajnikant V.

AU - Hoffman, Andrew K.

AU - Wharry, Janelle P.

AU - Rebak, Raul B.

N1 - Funding Information: The financial support of Global Nuclear Fuel Americas, GE Hitachi Nuclear Energy, and GE Research is gratefully acknowledged. This material is based upon work supported by the Department of Energy under Award Number DE-NE0009047 , and the Department of Energy-Nuclear Science User Facilities program under award NSUF-RTE-20-4108 . Publisher Copyright: © 2023

PY - 2023/12/15

Y1 - 2023/12/15

N2 - This article investigates the fuel-cladding chemical interaction (FCCI) behavior of two commercial FeCrAl alloys, APMT composition (Fe-21Cr-5Al-3Mo wt.%) and C35M (Fe-13Cr-5Al-2Mo-0.2Si-0.03Y wt.%), after neutron irradiation. “H-cup” diffusion multiples of FeCrAl alloys and ceramic UO2 fuel were irradiated at a temperature of ∼300 °C to a total estimated burnup of 26 GWd/tHM. Post-irradiation Examination results demonstrate the excellent degradation resistance of FeCrAl alloys as accident tolerant fuel (ATF) cladding materials in light water reactor conditions. The study concludes that there was no irradiation-induced defects observed in either of the two commercial FeCrAl claddings. The formation of amorphous Al/U mixed oxide was observed at the fuel-clad interface, which can serve as a tritium permeation barrier and protect against potential chemical attack from the fuel. The study attributed the formation of amorphous Al/U mixed oxide to the low temperature and limited time of neutron irradiation. APMT forms more distinct Cr and Cr-Fe intermetallic at the FeCrAl-UO2 interface than C35M due to the higher bulk Cr:Al ratio.

AB - This article investigates the fuel-cladding chemical interaction (FCCI) behavior of two commercial FeCrAl alloys, APMT composition (Fe-21Cr-5Al-3Mo wt.%) and C35M (Fe-13Cr-5Al-2Mo-0.2Si-0.03Y wt.%), after neutron irradiation. “H-cup” diffusion multiples of FeCrAl alloys and ceramic UO2 fuel were irradiated at a temperature of ∼300 °C to a total estimated burnup of 26 GWd/tHM. Post-irradiation Examination results demonstrate the excellent degradation resistance of FeCrAl alloys as accident tolerant fuel (ATF) cladding materials in light water reactor conditions. The study concludes that there was no irradiation-induced defects observed in either of the two commercial FeCrAl claddings. The formation of amorphous Al/U mixed oxide was observed at the fuel-clad interface, which can serve as a tritium permeation barrier and protect against potential chemical attack from the fuel. The study attributed the formation of amorphous Al/U mixed oxide to the low temperature and limited time of neutron irradiation. APMT forms more distinct Cr and Cr-Fe intermetallic at the FeCrAl-UO2 interface than C35M due to the higher bulk Cr:Al ratio.

KW - Accident tolerant fuel cladding

KW - FeCrAl alloy

KW - Fuel clad chemical interaction

UR - https://www.scopus.com/pages/publications/85170278607

UR - https://www.mendeley.com/catalogue/08d83db1-52eb-33a0-b056-750ff9ce40ed/

U2 - 10.1016/j.jnucmat.2023.154717

DO - 10.1016/j.jnucmat.2023.154717

M3 - Article

AN - SCOPUS:85170278607

SN - 0022-3115

VL - 587

JO - Journal of Nuclear Materials

JF - Journal of Nuclear Materials

M1 - 154717

ER -

FeCrAl fuel/clad chemical interaction in light water reactor environments

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Keywords

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